热中子俘获截面测量

用活化法测量了＾７１Ｇａ（ｎ,γ）＾７２Ｇａ、＾９４Ｚｒ（ｎ,γ）＾９５Ｚｒ、＾１９１Ｉｒ（ｎ,γ）＾１９２Ｉｒ＾ｍ１＋ｇ,ｍ２反应的热中子俘获截面,给出了最终结果,并与国外其他家的实验数据作了比较。     

Measurements of Neutron Capture Cross Section of 237Np for Fast Neutrons

The neutron capture cross section of ²³⁷Np has been measured for fast neutrons supplied at the center of the core in the Yayoi reactor. The activation method was used for the measurement, in which the amount of the product ²³⁸Np was determined by γ-ray spectroscopy using a Ge detector. The neutron flux at the center of the core calculated by the Monte Carlo simulation code MCNP was renormalized by using the activity of a gold activation foil irradiated simultaneously. The new convention is proposed in this paper to make possible a definite comparison of the integral measurement by the activation method using fast reactor neutrons with differential measurements using accelerator-based neutrons. “Representative neutron energy” is defined in the convention at which the cross section deduced by the activation measurement has a high sensitivity. The capture cross section of ²³⁷Np corresponding to the representative neutron energy was deduced as 0:80 ± 0:04b at 214 ± 9 keV from the measured reaction rate and the energy dependence of the cross section in the nuclear data library ENDF/B-VII.0. The deduced cross section of ²³⁷Np at the representative neutron energy agrees with the evaluated data of ENDF/B-VII.0, but is 15% higher than that of JENDL-3.3 and 13% higher than that of JENDL/AC-2008.     

Measurement of Thermal Neutron Cross Section of ^135Cs

350keV高压倍加器辐照装置的真空系统自2005年1月安装完成后，到目前已运行12个月，其运行状况正常。     

Isotopic Enrichment and Thermal Neutron Cross Section Measurement of Erbium-167

The thermal neutron cross section of Er167 has been obtained by measuring the relative number of Erl68 atoms formed by radiative neutron capture, when isotopically pure Er167 samples were exposed to a highly-thermalized neutron flux. A four-stage mass spectrometer was used to provide a 106/1 isotopic enrichment of Er167. It was also employed it measure the postirradiation Er168/Er167 isotopic ratios of the samples, which were placed in the D2O moderated Argonne CP-5 reactor. The integrated neutron flux, and effective neutron temperature in this reactor were also monitored mass spectrometrically, by observing isotopic ratio changes in nuclides whose cross section as a function of temperature are known. Two rare earth isotopes Sm149 and Gd157, were selected as the "non-l/v" absorbers that yielded this data. The effective neutron cross section of Er167 was found to be 740 ± 21 barns, at an effective neutron temperature of 42.8 + 2.5°C, with an integrated neutron flux of 3.37 ± 0.08 × 1019 neutron/cm2. The thermal (2200 m/sec) neutron cross section of Er167 was then calculated to be 699 ± 20 barns.     

Discussion of Data Difference Among Measured Thermal Neutron Cross Section of ^135Cs

“天光一号”系统采用诱导空间非相干（FEISI）和像传递技术实现了对靶平面的均匀辐照，但该技术并不能确保光束在激光传输路径上的近场分布均匀。虽然MOPA系统中大多数光学元件表面平均光通量不大于 1J／cm^2，     

Resonance Self-Shielding Corrections for Activation Cross Section Measurements

The Pade approximations of the Doppler broadening function ψ(θ, x) have been used for the calculations of resonance self-shielding factors used in activation measurements. It is shown that this method of the calculations is effective from the point of view of fastness and accuracy.     

Neutron Cross Section Covariances for Structural Materials and Fission Products

We describe neutron cross section covariances for 78 structural materials and fission products produced for the new US evaluated nuclear reaction library ENDF/B-VII.1. Neutron incident energies cover full range from ¹⁰−5 eV to 20 MeV and covariances are primarily provided for capture, elastic and inelastic scattering as well as (n,2n). The list of materials follows priorities defined by the Advanced Fuel Cycle Initiative, the major application being data adjustment for advanced fast reactor systems. Thus, in addition to 28 structural materials and 49 fission products, the list includes also ²³Na which is important fast reactor coolant. Due to extensive amount of materials, we adopted a variety of methodologies depending on the priority of a specific material. In the resolved resonance region we primarily used resonance parameter uncertainties given in Atlas of Neutron Resonances and either applied the kernel approximation to propagate these uncertainties into cross section uncertainties or resorted to simplified estimates based on integral quantities. For several priority materials we adopted MF32 covariances produced by SAMMY at ORNL, modified by us by adding MF33 covariances to account for systematic uncertainties. In the fast neutron region we resorted to three methods. The most sophisticated was EMPIRE-KALMAN method which combines experimental data from EXFOR library with nuclear reaction modeling and least-squares fitting. The two other methods used simplified estimates, either based on the propagation of nuclear reaction model parameter uncertainties or on a dispersion analysis of central cross section values in recent evaluated data files. All covariances were subject to quality assurance procedures adopted recently by CSEWG. In addition, tools were developed to allow inspection of processed covariances and computed integral quantities, and for comparing these values to data from the Atlas and the astrophysics database KADoNiS.     

中子核截面在线多普勒展宽方法研究

由中子截面多普勒展宽带来的反应性温度效应对反应堆中子学计算结果具有重要影响。基于自由气体模型和对靶核速度随机抽样的在线多普勒展宽方法,可使用0 K温度下的中子截面对给定温度的问题进行蒙特卡罗计算,摆脱对专用多普勒展宽程序的依赖。本文通过对在线多普勒展宽方法的程序实现,针对典型算例进行了验证和分析,证明了该方法能处理反应性温度效应,并对其适用性和未来发展前景进行了评价。     

热能点裂变截面系统规律

为建立缺乏实验测量数据的热能点裂变截面的评价方法和手段,本文从裂变理论基本公式出发,提出了其系统学公式的基本形式。在推荐热能点裂变截面以及单峰裂变位垒高度的基础上,研究了热能点裂变截面与鞍点态激发能之间的关系,发现了核素的热能点裂变截面与激发能之间的对数关系,通过拟合得到了质子数Z对应的系统学参数。进一步研究这些参数随Z的变化规律,发现了参数与Z的奇偶的关系,并通过拟合建立了全局的系统学公式。研究结果表明,热能点裂变截面与Z相关。     

235U快中子核反应截面协方差评价

本文基于最小二乘不确定度传递方法,建立²³⁵U中子裂变核反应截面模型依赖型与非模型依赖型协方差评价体系。通过针对实验测量较丰富的中子反应总截面、辐射俘获、(n,2n)等核反应实验数据不确定度源项分析,为协方差评价提供实验基础,并给出对应核反应截面的非模型依赖型协方差评价数据。通过开展快中子能区²³⁵U核反应理论模型参数灵敏度计算与分析,导出实验测量缺乏的核反应截面模型依赖型协方差评价数据。经上述系统评价,所得协方差数据与核反应截面中心值研究过程自洽、物理合理,并按国际标准ENDF-6格式输出,便于核工程用户使用。     

Neutron Activation Cross Section of Molybdenum Isotopes at 14.8 MeV

The neutron activation cross sections of Mo isotopes have been measured for the 14.8 MeV neutron. The cross sections have been determined with reference to the known ²⁷A1 (n, α)²⁴Na and the ²⁷Al(n, p)²⁷Mg reactions. The cyclic activation method was employed for the γ-ray measurement of short-lived nuclei. A 55 cm³ Ge(Li) detector was used for the measurement of γ-ray spectra. Cross section data are presented for (n, 2n), (n, p) and (n, a) reactions on Mo isotopes. The cross sections of (n, np) reactions on ⁹⁸Mo are also presented. The exponential dependence on (N-Z)/A of the (n, p) reaction cross sections are discussed.     

Reactor Neutron Capture Cross Section of 60.3-day 124Sb

The radioactive nuclides ¹²⁴Sb (T _1/2=60.3d) and ¹²⁵Sb (T_1/2=2.77yr) were produced from natural antimony by JRR-3 reactor irradiation of 283.5 h through the single and double capture processes. After cooling of 3.50 yr, the γ-ray spectrum of the antimony sample irradiated was measured by a 50 cc coaxial type Ge(Li) detector, and the photo-peak yield ratio of ¹²⁵Sb (E _r=428keV) to ¹²⁴Sb (E _r=1.691 MeV) was obtained. By using a relation between this photo-peak yield ratio and the ¹²⁴Sb (n, γ) ¹²⁶Sb cross section, the reactor neutron capture cross section of 60.3-day ¹²⁴Sb was obtained as 17.4:5:^+2.8 _?2.5b. The thermal neutron flux at the position of antimony sample irradiated was estimated as (4.92±0.38) ×10¹²n/cm²·s by measuring the 1.333-MeV photo-peak yield of ⁶⁰Co, which was activated by reactor irradiation of cobalt impurity contained in the antimony sample.     

Neutron Cross Section and Emission Spectra Calculations of Na Nucleus

Flibe (Li₂BeF₄) and Flinabe (LiFNaFBeF₂) are molten salt which are used as a first wall and blanket in advanced power extraction. Since they have low vapor pressure at high temperature and don’t pollute plasma. Moreover, they are used as a coolant material due to their high thermal cycling capacity. There are lots of benefits of using liquid wall such as reducing cost, increasing the reactor power and extending the life time of first steel wall. Being in Flinabe (LiFNaFBeF₂) molten salt, ²³ Na is also considered as a coolant material in fast reactors. Therefore, the reaction including neutron and ²³ Na is getting important. In this study, the cross section calculations of neutron induced reactions for ²³ Na were performed. Using pre-equilibrium and equilibrium models, excitation functions of ²³ Na(n,p)²³ Ne, ²³ Na(n,d)²² Ne, ²³ Na(n,α)²⁰ F, ²³ Na(n,2n)²² Na reactions and ²³ Na(n,xα), ²³ Na(n,xn) emission spectrums were calculated. We calculated alpha emission spectrum obtained when ²³ Na nuclei projectiled with 14.1 MeV neutrons and also neutron emission spectrum emitted when ²³ Na nuclei was bombardment with 14.6 MeV neutrons.     

Neutron Total Cross Section for Iron around Resonance Valley near 24 keV

A neutron leakage spectrum from a nickel sphere surrounding a 14-MeV neutron source is measured and analyzed in order to verify the accuracy of nickel cross sections. Measurement is done by means of a time-of-flight technique in the range of 2–15 MeV using an NE213 scintillator, and compared with calculations carried out with MCNP, a continuous energy Monte Carlo transport code, using JENDL-3PR1 and ENDFIB-IV neutron libraries.

In spite of an overestimation of neutron flux near 13 MeV, the calculated result employing JENDL-3PR1 shows generally better agreement with the measured spectrum. In ENDF/B-IV usage, there is disagreement between measured and calculated spectra between 5 and 12 MeV. Problems in evaluated nuclear data for nickel are also described.     

Radiative Neutron Capture Cross Section Measurement in Resonance Energy Region

<正>Radiative neutron capture cross-section (σnγ)is one kind of the nuclear data.The neutron reaction cross sections change sharply with the neutron energy in the resonance energy region,and this cross section can only be determined by experimental measurement.A C6D6 detector system which is used for the measurement of theσ_(nγ) has been installed on the Back-n of the China Spallation Neutron Source.The measurements for theσ_(nγ) of~(197) Au,